Samim Anghaie

THE ENHANCEMENTS AND TESTING FOR THE MCNPX 2.6.0 DEPLETION CAPABILITY

Monte Carlo–linked depletion methods have gained recent interest due to the ability to model complex three-dimensional geometries using continuous-energy cross sections. The integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a completely self-contained Monte Carlo-linked depletion capability in a single Monte Carlo code that is compatible with most nuclear criticality (KCODE) particle tracking features in MCNPX. MCNPX depletion tracks all necessary reaction rates and follows as many isotopes as cross-section data permit. The objective of this work is (a) describe the MCNPX depletion methodology dating from the original linking of MONTEBURNS and MCNP to the first public release of the integrated capability (MCNPX 2.6.B, June 2006) that has been reported previously, (b) further detail the many new depletion capability enhancements since then leading to the present Radiation Safety Information Computational Center (RSICC) release, MCNPX 2.6.0, (c) report calculation results for the H. B. Robinson benchmark, and (d) detail new features available in MCNPX 2.7.A. Each version of MCNPX depletion starting from MCNPX 2.6.A leading to the official RSICC release of MCNPX 2.6.0 and the new beta release MCNPX 2.7.A included significant upgrades that addressed key issues from earlier versions. This paper details these key issues and the approach utilized to address the issues as enhancements for MCNPX 2.6.0. The MCNPX 2.6.0 depletion capability enhancements include (a) allowing the modeling of as large a system as computer memory capacity permits; (b) tracking every fission product available in ENDF/B VII.0; (c) enabling depletion in repeated structures geometries such as repeated arrays of fuel pins; (d) including metastable isotopes in burnup; and (e) manually changing the concentrations of any isotope during any time step by specified atom fraction, weight fraction, atom density, or gram density. These enhancements allow better detail to model the true system physics as well as to improve the robustness of the capability. H. B. Robinson benchmark calculations were completed to assess the validity of nuclide predictability of MCNPX 2.6.0. The results show comparisons of key actinide and fission products as compared to experiment and the SCALE-4 SAS2H sequence 27-group cross-section library (27BURNUPLIB) results. MCNPX 2.6.0 depletion results are within 4% of the experimental results for most major actinides. Two major depletion enhancements are available in the MCNPX 2.7.A beta release: improved 63-group flux querying and parallelization of the burnup interface routines in multiprocessor mode. Fixing the energy group querying routine does correctly tally the energy flux for use with isotopes not containing transport cross sections; however, results show <1% change in nuclide prediction for the benchmark test case. MCNPX 2.7.A parallelizes the depletion interface routines and running of CINDER90 so that different burnable regions of a given depletion system can be preprocessed, burned, and postprocessed on separate slave processors. The parallelization involves minimal communication between processors and therefore leads to significant computational performance enhancement. The combination of new enhancements and testing of the MCNPX 2.6.0 depletion computational system make this capability a valuable Monte Carlo-linked depletion tool. Additional testing and feature enhancements are under development to further improve the usefulness of the computational tool.

Processing and fabrication of mixed uranium/refractory metal carbide fuels with liquid-phase sintering

Abstract Optimization of powder processing techniques were sought for the fabrication of single-phase, solid-solution mixed uranium/refractory metal carbide nuclear fuels – namely (U, Zr, Nb)C. These advanced, ultra-high temperature nuclear fuels have great potential for improved performance over graphite matrix, dispersed fuels tested in the Rover/NERVA program of the 1960s and early 1970s. Hypostoichiometric fuel samples with carbon-to-metal ratios of 0.98, uranium metal mole fractions of 5% and 10%, and porosities less than 5% were fabricated. These qualities should provide for the longest life and highest performance capability for these fuels. Study and optimization of processing methods were necessary to provide the quality assurance of samples for meaningful testing and assessment of performance for nuclear thermal propulsion applications. The processing parameters and benefits of enhanced sintering by uranium carbide liquid-phase sintering were established for the rapid and effective consolidation and formation of a solid-solution mixed carbide nuclear fuel.

An experimental study of electrical impedance tomography for the two-phase flow visualization

Abstract Numerical and experimental works are conducted to develop a visualization technique for the phase distribution in a two-phase flow field with electrical impedance tomography technique, which reconstructs the resistivity distribution with the electrical responses that are determined by corresponding excitations. The finite element method is employed to solve the electrical field induced by the currents through electrodes placed along the boundary and a regularized Newton-Raphson iterative method is used to determine the search step minimizing the error between the calculated and the measured voltages at the electrodes. With an apparatus developed for impedance imaging this study attempts to reconstruct the images of the simulated bubble distributions and the reconstructed images imply the potential possibility of the electrical impedance tomography for the two-phase flow visualization.

Thermal hydraulic design analysis of ternary carbide fueled square-lattice honeycomb nuclear rocket engine

A computational analysis is conducted to determine the optimum thermal-hydraulic design parameters for a square-lattice honeycomb nuclear rocket engine core that will incorporate ternary carbide based uranium fuels. Recent studies at the Innovative Nuclear Space Power and Propulsion Institute (INSPI) have demonstrated the feasibility of processing solid solution, ternary carbide fuels such as (U, Zr, Nb)C, (U, Zr, Ta)C, (U, Zr, Hf)C and (U, Zr, W)C. The square-lattice honeycomb design provides high strength and is amenable to the processing complexities of these ultrahigh temperature fuels. A parametric analysis is conducted to examine how core geometry, fuel thickness and the propellant flow area effect the thermal performance of the nuclear rocket engine. The principal variables include core size (length and diameter) and fuel element dimensions. The optimum core configuration requires a balance between high specific impulse and thrust level performance, and maintaining the temperature and strength limi...

A Computational Fluid Dynamics and Heat Transfer Model for Gaseous Core and Very High Temperature Gas-Cooled Reactors

A computational approach to the solution of Navier-Stokes equations for the thermal and flow fields of very high temperature gas-cooled and gaseous core reactors is presented. An implicit-explicit, finite volume, MacCormack method, in conjunction with the Gauss-Seidel line iteration procedure, is utilized to solve axisymmetric, thin-layer Navier-Stokes equations. An enthalpy rebalancing scheme is implemented to allow the convergence solutions to be obtained with the application of a wall heat flux. The subsonic and supersonic flows of helium in a very high temperature gas-cooled reactor and uranium tetrafluoride (UF 4 ) in a gaseous core reactor under variable boundary conditions (such as adiabatic, isothermal, and constant heat flux) are calculated. The numerical results are compared with other published results and experimental-based correlations. The good agreement with empirical correlations indicates the usefulness of the presented model for the prediction of the flow and temperature distribution under the convective and radiative heat transfer environment of very high temperature gas-cooled and gaseous core reactors.

Improved Reaction Rate Tracking and Fission Product Yield Determinations for the Monte Carlo-Linked Depletion Capability in MCNPX

Abstract As advanced reactor concepts challenge the accuracy of current modeling technologies, a higher-fidelity depletion calculation is necessary to model time-dependent core reactivity properly for accurate cycle length and safety margin determinations. The recent integration of CINDER90 into the MCNPX Monte Carlo radiation transport code provides a completely self-contained Monte Carlo-linked depletion capability. Two advances have been made in the latest MCNPX capability based on problems observed in prereleased versions: continuous-energy collision density tracking and adequate fission yield selection. Prereleased versions of the MCNPX depletion code calculated the reaction rates for (n,2n), (n,3n), (n,p), and (n,α) by matching the MCNPX steady-state 63-group flux with 63-group cross sections inherent in the CINDER90 library and then collapsing to one-group collision densities for the depletion calculation. The accuracy of this procedure is therefore dictated by the adequacy of the 63-group energy structure of the cross-section set to accurately capture the spectrum of a specific model. Different types of models would therefore require different types of cross-section energy group structure. MCNPX 2.6.A eliminates this dependency by using the continuous-energy reaction rates determined during the MCNPX steady-state calculation to calculate energy-integrated collision rates to be used by CINDER90. MCNPX 2.6.A now also determines the proper fission yield to be used by the CINDER90 code for the depletion calculation. The CINDER90 code offers a thermal, fast, and high-energy fission yield for each fissile isotope contained in the CINDER90 data file. MCNPX 2.6.A determines which fission yield to use for a specified problem by calculating the integral fission rate for the defined energy boundaries (thermal, fast, and high energy), determining which energy range contains the majority of fissions, and then selecting the appropriate fission yield for the energy range containing the majority of fissions. The MCNPX depletion capability enables complete, relatively easy-to-use depletion calculations in a single Monte Carlo code. This study focuses on the methodology development of the two improvements described here. Further improvements are under development to enhance the usefulness of this new capability.

Multimegawatt NEP with vapor core reactor MHD

Efforts at the Innovative Nuclear Space Power and Propulsion Institute have assessed the feasibility of combining gaseous or vapor core reactors with magnetohydrodynamic power generators to provide extremely high quality, high density, and low specific mass electrical power for space applications. Innovative shielding strategies are employed to maintain an effective but relatively low mass shield, which is the most dominating part of multi-megawatt space power systems. The fission driven magnetohydrodynamic generator produces tens of kilowatt DC power at specific mass of less than 0.5 kg/kW for the total power system. The MHD output with minor conditioning is coupled to magnetoplasmadynamic thruster to achieve an overall NEP system specific mass of less than 1.0 kg/kW for power levels above 20 MWe. Few other concepts would allow comparable ensuing payload savings and flexible mission abort options for manned flights to Mars for example.

An effective conduction length model in the enthalpy formulation for the stefan problem

Abstract In the fixed-grid finite-volume formulation, so called the enthalpy formulation, for the Stefan problem, the temperature and the front movement show step-like history, which is a well-known characteristic of the enthalpy method. This paper presents an effective conduction length model to mitigate such an oscillatory behavior as well as to support the physical reasoning. The proposed model is based on the simple fact that the heat flux across the boundary of phase-change cells should be estimated with the distance between the phase front and the center of neighboring cell. The model is applied to one-dimensional Stefan problems with various Stefan numbers. The numerical results show that the proposed model can smooth the spurious oscillation of the history of temperature and the evolution of front movement.

Thermal-hydraulic analysis of bulk evaporation and condensation in a multiphase nuclear fuel cell

A thermal-hydraulic model is developed to simulate and study the dynamic behavior of bulk evaporation and condensation processes in a multiphase nuclear fuel cell. The phase-change process is driven and controlled by internal heat generation and wall heat removal under constant volume condition. The modeling involves variable gravity conditions that allow for performance analysis of the multiphase nuclear fuel for terrestrial and space applications. A complete set of governing equations for both liquid and vapor phases is developed and numerically solved. The model is used to simulate the operation of a multiphase nuclear fuel cell at zero-gravity and microgravity levels. The temperature and phase distribution, the flow field, and the evolution of the liquid-vapor interface are computed and demonstrated.

Nuclear design analysis of square-lattice honeycomb space nuclear rocket engine

The square-lattice honeycomb reactor is designed based on a cylindrical core that is determined to have critical diameter and length of 0.50 m and 0.50 c, respectively. A 0.10-cm thick radial graphite reflector, in addition to a 0.20-m thick axial graphite reflector are used to reduce neutron leakage from the reactor. The core is fueled with solid solution of 93% enriched (U, Zr, Nb)C, which is one of several ternary uranium carbides that are considered for this concept. The fuel is to be fabricated as 2 mm grooved (U, Zr, Nb)C wafers. The fuel wafers are used to form square-lattice honeycomb fuel assemblies, 0.10 m in length with 30% cross-sectional flow area. Five fuel assemblies are stacked up axially to form the reactor core. Based on the 30% void fraction, the width of the square flow channel is about 1.3 mm. The hydrogen propellant is passed through these flow channels and removes the heat from the reactor core. To perform nuclear design analysis, a series of neutron transport and diffusion codes ar...